Disposal of radioactive wastes has become an increasingly difficult task presenting serious impediments to the development and utilization of nuclear power facilities as an alternate source of energy. Large quantities of toxic materials such as high level radioactive wastes are often stored in spent reactor storage pools, or generated during reprocessing of spent power reactor fuel or in the operation and maintenance of nuclear power plants. These radioactive wastes must be disposed of safely and efficiently.
The difficulty in the disposal of radioactive waste is even more acute when the concentration of the radioactive moieties exceed about one microcurie per cubic centimeter of waste stream and the radioactive species have a multiyear half-life necessitating immobilization of the radioactive moities for a period greater than a century.
One of the most generally accepted procedures for the disposal of such wastes is to convert the radioactive waste to dry solid form so that the waste is rendered chemically, thermally and radiolytically stable.
Organic ion exchange media have been employed for the removal of the radioactive ions contained in such wastes but, due to their low theshhold for radiation damage, they are not suitable for this purpose. The ultimate radioactive loading level of organic ion exchange media is more limited than inorganic ion exchange media since the former is susceptible to radiation damage at much lower dosage than the inorganic ion exchange media.
Among the inorganic ion exchange media which have been used in the removal of radioactive ions from radioactive wastes, those based on a porous glass matrix as described in copending application Ser. No. 039,595, filed May 16, 1979 and copending application Ser. No. 065,572 filed Aug. 10, 1979, have proven to be most useful. Zeolites and sodium titanates have also achieved some degree of acceptability as inorganic ion exchange media for this purpose.
In general, the method employed for the removal of the radioactive ions from liquids (e.g., radioactive waste stream) containing the same comprises passing the waste stream through a suitable container (preferably made of stainless steel) containing the ion exchange medium until the ion exchange capacity of the medium is essentially used. Once the ion exchange capacity of the medium has been used, the problem becomes one of safe disposal of the container containing the spent medium without attendant radiation damage and hazards.
Heretofore, one such disposal procedure for the so-called "low level" radioactive waste streams involved draining the container, sealing and burying the same underground in shallow sites (usually less than about six meters deep).
If the container is improperly sealed, there is the likelihood, and danger, that during trnasportation to the burial site, the seal may break, causing radioactive material to be scattered into the surrounding areas. In addition, and in areas having extensive rainfall, the steel container will, with passage of a relatively short period of time, rust and thus expose the spent mass to water which leaches the radioactive ions, especially Cs.sup.137, Cs.sup.134, Sr.sup.90, Co.sup.60, etc. Consequently, the ground water will become contaminated with these ions.
Another method of disposing of the spent ion exchange media involves their solidification with bitumen. However, this procedure causes radiation damage to the bitumen and is a fire hazard.
The spent ion exchange media have also been disposed of by mixing with cement to form concrete and burying the resulting mass. As in the burial method heretofore described, however, this procedure also presents the problem associated with leaching of the radioactive ions and contamination of the ground water.
The ion exchange media can be placed in a high integrity container designed to last several hundred years. However, during this period, the ion exchange media will release gases due to, for example, radiolosis and decay of organic matter. If this gas is not vented, the container will explode. If it is vented, ground water will enter the container. In either case, it will not perform as a high integrity container.
Another method of disposal of spent radioactive ion exchange medium in a loose, dry-powdered form involves removal of said medium from the ion-exchange container and melting it with glass frit to form borosilicate glass. The requirement of handling such loose, dry powders which are not only abrasive but also have a significant amount of respirable fines, dictates the use of hot cells, complicated out gassing systems and remote maintenance, all of which translates into an expensive operation. See, Electric Power Research Institute Report Number EPRI NP-1087 SIA 78-414 "Nuclear Waste Management Status and Recent Accomplishments" May, 1979.
Thus, notwithstanding numerous methods which have heretofore been proposed for radioactive wastes disposal, and a variety of compositions employed for decontaminating such wastes, the problem of effective decontamination of these streams and safe disposal of the spent media resulting from the decontamination process continues to present serious difficulties. Consequently, effective utilization of nuclear power plant facilities still remains dependent on the development of safe and environmentally acceptable methods of decontamination of radioactive waste streams and the disposal of the spent ion exchange mass resulting from treatment of these streams.
Accordingly, an object of this invention is to provide a novel method of removing radioactive ions from a liquid containing the same which comprises contacting said liquid with a novel composition having ion-exchange capability (which removes said radioactive ions from said liquid), followed by sintering the resulting radioactive composition below the temperature which causes substantial volatilization and escape of the radioactive species to the atmosphere.
Another object of the invention is to provide a novel method of decontaminating a radioactive-containing liquid which comprises contacting said liquid with an inorganic ion-exchange material characterized by a relatively high sintering temperature and a high capacity for radioactive species, there-after adding to the resulting contaminated material an additive that has the effect of lowering the sintering temperature, followed by sintering the resulting admixture without disseminating volatilized radioactive species to the atmosphere.
Another object of the invention is to provide a novel composition comprising high melting, inorganic ion-exchange material admixed with a sufficient amount of an additive, the resulting admixture having a sintering temperature significantly lower than the ion-exchange material per se.
It is another object of this invention to provide a novel method of making the aforesaid inorganic ion-exchange admixtures, in particular, homogeneous admixtures.
Another object of the invention is directed to novel articles of manufacture, in particular, novel containers capable of withstanding high sintering temperatures and which include with their structure the aforesaid novel inorganic ion-exchange admixtures.
It is another object of the invention to provide novel compositions comprising radioactive inorganic ion-exchange material plus an additive hereinafter defined, said compositions characterized by relatively low sintering temperatures whereby substantial dissemination of the radioactive species in said material is prevented during sintering of the compositions.
It is also an object of this invention to provide a novel apparatus for removing radioactive species from a nuclear waste stream, said apparatus including within its structure the aforesaid novel compositions which can be sintered within the environment of said waste stream.